• 전자공학회논문지 IE
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• v.45 no.2
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• pp.6-12
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• 2008

In this study, because of the importance of asphalt content, the target is to develop a asphalt content detector that the influence of the jurisdiction and usage of radioisotopes are limited of 100[${\mu}Ci$] or less. For development of a asphalt content detector, it is divided into three parts that is affected the change of asphalt content due to the position of neutron source. First part is to decrease the distance between asphalt mixture and neutron source, second is to install the reflector, and last is to change the history number. It performed to analysis by computer simulation and satisfactory results of error range can use to design data for development of a asphalt content detector.

• Journal of Radiation Protection and Research
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• v.43 no.4
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• pp.131-136
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• 2018

Background: We investigated the current characteristics of a thin-film Ag electrode on a chemical vapor deposition (CVD) diamond. The CVD diamond is widely recognized as a radiation detection material because of its high tolerance against high radiation, stable response to various dose rates, and good sensitivity. Additionally, thin-film Ag has been widely used as an electrode with high electrical conductivity. Materials and Methods: Considering these properties, the thin-film Ag electrode was deposited onto CVD diamonds with varied deposition thicknesses (${\fallingdotseq}50/98/152/257nm$); subsequently, the surface thickness, surface roughness, leakage current, and photo-current were characterized. Results and Discussion: The leakage current was found to be very low, and the photo-current output signal was observed as stable for a deposited film thickness of 98 nm; at this thickness, a uniform and constant surface roughness of the deposited thin-film Ag electrode were obtained. Conclusion: We found that a CVD diamond radiation detector with a thin-film Ag electrode deposition thickness close to 100 nm exhibited minimal leakage current and yielded a highly stable output signal.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1398-1406
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• 2024

The control rod worth is a key safety parameter required to be measured in commercial pressurized water reactors (PWRs). Conventionally, the control rod worth is measured after reaching the critical state, which occupies the considerable time in the zero-power physics test. In this study, an efficient control-rod worth measurement technique has been proposed based on the improved neutron-source multiplication method, which can be implemented with the source-range detector count rates in the subcritical states. Moreover, the noise reduction technique has been adopted to smooth the large fluctuation existing in the original signals. In order to verify the engineering performance of the proposed measurement technique, the measured source-range detector count rates during the rod withdrawal process before reaching critical state in a CNP1000 reactor have been employed. It demonstrated that almost all estimated results of control rod worth satisfy the engineering acceptance criteria, except one control rod with the relative difference over 10 %, which indicates the capability of the proposed method in estimating control rod worth.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1354-1357
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• 2017

Characterization of special nuclear material may be performed using energy spectroscopy of either the neutron or gamma-ray emissions from the sample. Gamma-ray spectroscopy can be performed relatively easily using high-resolution semiconductors such as high-purity germanium. Neutron spectroscopy, by contrast, is a complex inverse problem. Here, results are presented for $^{252}Cf$ and PuBe energy spectra unfolded using a single EJ309 organic scintillator; excellent agreement is observed with the reference spectra. Neutron energy spectroscopy is also possible using a two-plane detector array, whereby time-of-flight kinematics can be used. With this system, energy spectra can also be obtained as a function of position. Spatial-dependent energy spectra are presented for neutron and gamma-ray sources that are in excellent agreement with expectations.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.50-50
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• 1999

• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.334-338
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• 1973

The horizontal distribution of the fast neutron flux density in the Bulk Shielding Experimental Tank of the TRIGA Mark-II reactor at the steady power of 250 KW has been measured using a solid state track detector which is natural mica placed in contact with $^{232}$ Th fissile foil. The neutron flux density was calculated on the assumption that the fast neutron spectrum is similar to that from the thermal-induced $^{235}$ U fission. The resulting flux density distribution along the horizontal line from the center of the thermalizing column door is presented in tabular and graphical forms.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.203-210
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• 2018

As the modernization of the nuclear instrumentation system progresses, research reactors have adopted digital wide-range neutron power measurement (DWRNPM) systems. These systems typically monitor the neutron flux across a range of over 10 decades. Because neutron detectors only measure the local neutron flux at their position, the local neutron flux must be converted to total reactor power through calibration, which involves mapping the local neutron flux level to a reference reactor power. Conventionally, the neutron power range is divided into smaller subranges because the neutron detector signal characteristics and the reference reactor power estimation methods are different for each subrange. Therefore, many factors should be considered when preparing the calibration procedure for DWRNPM channels. The main purpose of this work is to serve as a reference for performing the calibration of DWRNPM systems in research reactors. This work provides a comprehensive overview of the calibration of DWRNPM channels by describing the configuration of the DWRNPM system and by summarizing the theories of operation and the reference power estimation methods with their associated calibration procedure. The calibration procedure was actually performed during the commissioning of an open-pool type research reactor, and the results and experience are documented herein.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1964-1969
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• 2021

We developed a Dual-PPACs detector for fast neutron measurements that consists of two sets of PPAC: conventional PPAC and fission PPAC. A²³⁸U(U₃O₈) coating is placed in the fission PPAC's anode, which is used as the neutrons conversion layer. An experiment was performed to measure neutron time-of-flight (TOF) in which ²⁵²Cf spontaneous fission source was used. An excellent time resolution of 164ps has been observed at 6 mbar in isobutene gas. With the excellent time resolution of Dual-PPACs detector, exact neutron energy can be extracted from the timing measurement. The experimental detection efficiency was 1.9 × 10^-7, consistent with the efficiency of 2.5 × 10^-7 given by a Geant4 simulation. Ultimately, the results show that the Dual-PPACs detector is a suitable candidate for measuring fast neutrons in the future CiADS system.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.273-280
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• 2014

The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software 'K-SWR'. The detectors' response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403). The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the $^{241}Am$-Be sources held in a graphite pile, a bare $^{241}Am$-Be source, and a DT neutron generator. Fluence-average energy ($E_{ave}$) varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [$H^*(10)/h$] varied from 0.99 to 16.5 mSv/h.